SOME NUCLEAR FACILITIES OF SSC RF-IPPE AND CHARACTERISTICS OF SPENT NUCLEAR FUEL

World's first NPP with AM reactor commissioned on June 27, 1954, is one of the Institute's experimental base objects. Its design-basis thermal power was 30 MW, and electrical one - 5 MW. After the initial 5-year period of operation in the NPP mode, it has been used as an experimental facility for testing various fuel elements and structural materials, testing and improvement of water-chemical modes, and as an irradiation device for radio-isotopic production build-up. The reactor core includes 128 fuel channels (FC), each containing 4 fuel elements (FE). Initially, uranium-molybdenum fuel was used (OM-9), subsequently replaced by uranium dioxide (UO2) with low enrichment. Characteristics of nuclear fuel are given in Table 1.

In 1956, BR-2 mercury-cooled, 100 kW power reactor was commissioned at IPPE. Principle feasibility of using fast neutron reactors for power generation was shown with that reactor. In 1958, the BR-2 was replaced with BR-5, sodium-cooled, 5 MW. That reactor was used for the development of engineering and technological solutions for their subsequent use in commercial fast neutron reactors designing and construction. Fuel elements of various compositions were used for that purpose , as shown in Table 2 (plutonium dioxide PuO2, uranium nitride UN, uranium carbide UC).

Burn-up levels of 6-7% have been achieved, experience of operating non-sealed fuel elements gained, release of fission products into the coolant and gas plenums of primary circuit has been studied; the system for monitoring of non-sealed FE for delayed neutrons is created; safety system and dynamic monitoring of the reactor facility has been investigated, with a series of neutronic and material irradiation studies accomplished.

In 1973, reconstruction of BR-5 reactor was completed. Its power was upgraded to 10 MW, and it has been termed BR-10 since (see Table 2). To date, reactor BR-10 has been used for the studies of nuclear fuel and materials performance, isotope production, and neutron therapy. Engineering options for upgrading safety of fast-neutron power reactors have been tested and optimized at the facility.

Table 1 Irradiation Parameters of AM reactor SA Notes: 1. The number of SAs in the reactor - 128;

2. The number of fuel rods per SA - 4;

3. Fuel composition length - 1700 mm 4. Fuel rod can thickness - (0.15-0.20) mm

") - According to the Russian classification of steels.

Table 2 Irradiation Parameters of BR (fast-neutron) reactor SA of the fuel part

(mm) Notes: 1. The number of SAs in the reactor - 128 (except for the Pu core);

2. Fuel composition length for the Pu core - 130 mm; others - 400 mm.

) - According to the Russian classification of steels.

Experimental transportable power plant TES-3 of 1.5 MW power was developed and commissioned at the Institute in 1961. Fuel elements for it were fabricated from 3 alloy with enrichment of uranium-235 of 2.4 to 3.6%, and OM-9 alloy with 36 to OM-90% enrichment of U-235. Diameter of fuel elements amounted to 12 mm, and the length of fuel composition - ~lm. One fuel subassembly (SA) contained ~1 kg of uranium. Average bum-up reached 3%. The experience gained with the TES-3 development was used in the development of unit-transportable and naval NPPs with two-circuit integral self-regulating water-cooled-water-moderated reactor facilities with natural circulation of coolant in primary and secondary circuits. After the test operation of NPP, TES-3 was decommissioned in 1978.

In 1970-1984, at the IPPE's special test facility was used for testing seven specimens of power facility with thermionic reactor-converter "TOPAZ" and pre-launch tests of two standard facilities. Power generating channels of the "TOPAZ"

were developed, fabricated, and undergone in-pile tests at the EPPE. Those FEs were fabricated from uranium dioxide with enrichment of U-235 ranging from 21%to 90%.

The fuel elements had diameters of 10 mm, with the length of fuel composition of-300 mm. The content of uranium per FE was 196g. Average burn-up amounted to 0.4%.

For reactor materials and fuel composition studies, the Institute has a complex of "hot" chambers and "heavy" boxes. Experimental and standard FE, SA, ampoules, blocks, items, and material samples received from research reactors AM, BR-10, BOR-60, and others, as well as from power reactor BN-350 are dismantled and investigated in these chambers and boxes.

4. TECHNOLOGY OF SPENT FUEL STORAGE AT SSC RF-IPPE

As a result of long-term operation of the experimental base nuclear facilities, a large quantity of spent nuclear materials has been accumulated at the Institute. They differ in their types (uranium, plutonium), enrichment of U-235 (ranging from 4% to 90%), burn-up depth, "cooling" period after irradiation, composition of the fuel contained, etc.

Spent fuel elements of research reactors are stored in a special storage designed for temporary storing of spent fuel (Central storage for spent fuel, CSSF). This storage is situated within the Institute's guarded perimeter with the status of protected area, remote from other production facilities. The CCSF belongs to the dry type storage; it has above and underground parts.

The upper part contains the storage life-support systems, such as: power- and water supply, special waste-water disposal, dosimetry control post, fire- and emergency alarm systems, ventilation, heating, and automatic water pumping-down devices. For spent nuclear fuel handling, the storage is equipped with overhead travelling crane of 20/5 t load-carrying capacity, reloading container, and various devices.

Spent fuel is kept in the underground part in special cells. The cells are made from heavy concrete; they form a square grid with the "pitch" of 800 mm. Stainless steel covers of different lengths are installed into the cells, with FE inserted therein.

Each cell is covered with protecting steel plug and rubber-sealed lead. The monolith cover from heavy concrete, protecting plugs, sealing leads, and the underground compartment foundation ensure the decrease of y-radiation to admissible levels, both inside the storage, and outside, in complete loading of the storage with spent fuel of 6 x 10⁷ g eq. radium activity.

Spent nuclear fuel received directly from the research reactors' cores is placed into the "cooling" ponds available at each research reactor, being stored there for 1.5 to 3 years, then dispatched to the CCSF. Therefore, a considerable diminishing of the activity and heat release from the spent FEs is achieved.

For the transportation of spent nuclear fuel from buildings it is handled in, from research reactors, "hot" chambers to the CCSF, the Institute' railway system is used, flat-car with removable protection container, hoisting devices, and diesel locomotive are used. Spent fuel is reloaded with remote control using a special reloading container.

The system of nuclear material accounting and control (MC&A) now in force at the Institute provides all necessary information on the content of each cell in the storage and cooling ponds of research reactors, as well as on each FE:

• Passport is made for each fuel element with reactor type indicated, as well as the manufacturer's serial number design specifications, mass of nuclear material before loading into reactor, and its calculated mass after unloading from the reactor, dates of loading and unloading, burn-up, number of the clad in which the FE is reloaded, date of dispatching for storage.

• Spent FE is transferred for storage after executing special documents signed by the Institute's managers;

• Special documents of CCSF record the following data for each cell of the storage: number, FE identities and quantity; passport numbers of the FE, and requirements for the FE transfer for storage, weight of uranium or plutonium; date of loading;

• After the clad with spent FEs is loaded into the cell, it is covered with sealing lead, with the date of loading and reactor type indicated on the lead;

• The FE presence is verified periodically by visual inspection of the storage cells;

• Concentrations of a-, p-active aerosols and y-radiation dose rates are checked at the CCSF periodically.

The question of further fate of the spent FEs stored at CCSF has been regarded very attentively at EPPE. The reasons for this keen interest are:

• The CCSF has been filled by 80%. The remaining free cells are not sufficient for the placement of spent FE stored at the cooling ponds at the research reactors AM, BR-10, and at the "hot" laboratory, and the fuel elements available at the AM and BR-10 at present.

• When designing and creating the research reactors, spent fuel problems were not considered, and as a result, technologies for spent nuclear fuel

regeneration are not available for all cases. The processing of spent fuel elements with low level of nuclear material in small batches proves to be economically unprofitable. The standard and experimental fuel elements stored at CCSF have large variations in both fuel composition and design. For example, the AM fuel compositions consist of both uranium dioxide with U-235 enrichment ranging from 2% to 10%, and the OM-9 alloy with enrichment of U-235 from 4.4% to 7.0%. As for the BR-10 fuel compositions, they include both plutonium dioxide, and uranium dioxide with enrichment of U-235 of 90%, and uranium nitride with that of 90%. Spent fuel elements received after investigations from the "hot" laboratory also have widely ranging fuel compositions with enrichment of U-235 of 4% to 90%.

• The storage for spent fuel now available was planned to be used for a temporary storage of such fuel, and in principle, for the time being it meets the requirements. However, with the issue of new regulatory documents that make these requirements more strict, the reconstruction and modernization of this storage has to be scheduled for the nearest time.

It should be pointed out that the current situation with spent nuclear fuel of research reactors at the Institute proves to be typical also for other facilities that have research reactors.

In the Declaration of the Moscow Summit of 8 leading nuclear states convened on the 19-th of April of 1996, it is stated (article 18) that "state bodies are obliged to ensure safe management of radioactive wastes, as well as the development of rules for their appropriate processing, storaging, and final disposal".

5. ANALYSIS OF SPENT NUCLEAR FUEL DURING LONG-TERM STORAGE